• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• 한국신재생에너지학회:학술대회논문집
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• 2010.06a
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• pp.228.2-228.2
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• 2010

In this work, the hydrogen production costs of the nuclear energy sources are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of HTE process with Very High Temperature nuclear Reactor (VHTR) as a thermal energy source rather than the LUEC (Levelized Unit Electricity Cost). The general ground rules and assumptions follow G4-ECONS. Through a preliminary study of cost estimates, we wished to evaluate the economic potential for hydrogen produced from nuclear energy, and, in addition, to promptly estimate the hydrogen production costs for an updated input data for capital costs. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if $CO_2$ capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of $CO_2$. Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1154-1162
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• 2019

China is undertaking an energy reform from fossil fuels to clean energy to accomplish $CO_2$ intensity (CI) reduction commitments. After hydropower, nuclear energy is potential based on breadthwise comparison with the world and analysis of government energy consumption (EC) plan. This paper establishes a CI energy policy response forecasting model based on national and provincial EC plans. This model is then applied in Fujian Province to predict its CI from 2016 to 2020. The result shows that CI declines at a range of 43%-53% compared to that in 2005 considering five conditions of economic growth in 2020. Furthermore, Fujian will achieve the national goals in advance because EC is controlled and nuclear energy ratio increased to 16.4% (the proportion of non-fossil in primary energy is 26.7%). Finally, the development of nuclear energy in China and the world are analyzed, and several policies for energy optimization and CI reduction are proposed.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.432-442
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• 2023

Characterizing the behavior of nuclear reactor plate fuels is vital to the progression of advanced fuel systems. The states of pre- and post-irradiation plates need to be determined effectively and efficiently prior to and following irradiation. Due to the hostile post-irradiation environment, characterization must be completed remotely. Laser-based characterization techniques enable the ability to make robust measurements inside a hot-cell environment. The Laser Shock (LS) technique generates high energy shockwaves that propagate through the plate and mechanically characterizes cladding-cladding interfaces. During an irradiation campaign, two Idaho National Laboratory (INL) fabricated MP-1 plates had a fuel breach in the cladding-cladding interface and trace amounts of fission products were released. The objective of this report is to characterize the cladding-cladding interface strengths in three plates fabricated using different fabrication processes. The goal is to assess the risk in irradiating future developmental and production fuel plates. Prior LS testing has shown weaker and more variability in bond strengths within INL MP-1 reference plates than in commercially produced vendor plates. Three fuel plates fabricated with different fabrication processes will be used to bound the bond strength threshold for plate irradiation insertion and assess the confidence of this threshold value.

• Analytical Science and Technology
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• v.14 no.4
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• pp.291-299
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• 2001

A method of separation and purification of Sm for quantitation of Sm isotopes from various fission products in PWR spent nuclear fuels has been studied. Simulated solution containing inactive metal ions(Cs, Ba, Gd, Eu, Sm and Nd) in place of radioactive fission products was prepared. Sm was separated with 0.5 M $HNO_3$/80% MeOH after washing with 1 M $HNO_3$/90% MeOH on AG $1{\times}8$, anion exchange resin. Sm was purified on cation exchange resin, AG $50W{\times}8$, pretreated with 0.2 M alpha-hydroxisobutyric acid(pH 4.5-4.6) to remove Ba causing isobaric effect Sm from PWR spent fuel. As a result of mass spectrometric measurement, eluted Sm portion did not include isobars form other elements such as Gd, Eu, Pm, Nd and BaO. The contents of Sm and its isotopes($^{147}Sm$, $^{148}Sm$, $^{149}Sm$, $^{150}Sm$, $^{151}Sm$, $^{152}Sm$ and $^{154}Sm$) in spent fuel were determined by isotope dilution mass spectrometric method spiking $^{154}Sm$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.65-72
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• 2008

The purpose of the HLW deep geological disposal is to isolate and to delay the radioactive material release to human beings and the environment for a long time so that the toxicity does not affect to the environment. The main requirements for the HLW repository design is to keep the buffer temperature below $100\;^{\circ}C$ in order to maintain its integrity. So the cooling time of spent fuels discharged from the nuclear power plant is the key consideration factors for efficiency and economic feasibility of the repository. The disposal tunnel/disposal hole spacing, the disposal area and thermal capacity required for the deep geological repository layout which satisfies the temperature requirement of the disposal system is analyzed to set the optimized spent fuels cooling time. To do this, based on the reference disposal concept, thermal stability analyses of the disposal system have been performed and the derived results have been compared by setting the spent fuels cooling time and the disposal tunnel/disposal hole spacing in various ways. From these results, desirable spent fuels cooling time in view of disposal area is derived. The results shows that the time reaching the maximum temperature within the design limit of the temperature in the disposal site is likely shortened as the cooling time of spent fuels becomes short. Also it seems that the temperature-rising and-dropping patterns in the disposal site are of smoothly varying form as the cooling time of spent fuels becomes long. In addition, it is revealed that a desirable cooling time of spent fuels is approximately 40-50 years when spent fuels are supposedly disposed in the deep geological disposal site with its structural scale under consideration in this study.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.247-258
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• 2011

Analysis of foreign and domestic patents for pyroprocessing technology of spent nuclear fuels was carried out in this study. The current status of pyroprocessing technology development in such countries as Korea, USA, Japan and EU was analyzed by classifying the patents for 1975 through 2009 according to registration country, assignee, calendar year and technology area. The major assignees' activity indices were compared in order to find out whether there is any concentrated area of technical details. Technology competitiveness of the countries was also investigated from the information of patent citation number and family size. Furthermore, some essential unit technologies required for the commercialization of pyroprocessing were derived and examined in the aspect of the state of art as well as the trend of technology development.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.29-40
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• 2013

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.223-228
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• 2018

AREVA NP (Courbevoie, Paris, France) is actively developing several enhanced accident-tolerant fuels cladding concepts ranging from near-term evolutionary (Cr-coated zirconium alloy cladding) to long-term revolutionary (SiC/SiC composite cladding) solutions, relying on its worldwide teams and partnerships, with programs and irradiations planned both in Europe and the United States. The most advanced and mature solution is a dense, adherent chromium coating on zirconium alloy cladding, which was initially developed along with the CEA and EDF in the French joint nuclear R&D program. The evaluation of the out-of-pile behavior of the Cr-coated cladding showed excellent results, suggesting enhanced reliability, enhanced operational flexibility, and improved economics in normal operating conditions. For example, because chromium is harder than zirconium, the Cr coating provides the cladding with a significantly improved wear resistance. Furthermore, Cr-coated samples exhibit extremely low corrosion kinetics in autoclave and prevents accelerated corrosion in harsh environments such as in water with 70 ppm Li leading to improved operational flexibility. Finally, AREVA NP has fabricated a physical vapor deposition prototype machine to coat full-length cladding tubes. This machine will be used for the manufacturing of full-length lead test rods in commercial reactors by 2019.